ML20080N331
| ML20080N331 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/01/1995 |
| From: | Bateman W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080N334 | List: |
| References | |
| NUDOCS 9503060231 | |
| Download: ML20080N331 (14) | |
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UNITED OTATES 7-
- J NUCLEAR REGULATORY COMMISSION '
WASHINGTON, D.C. SesEG eeM j
SOUTHERN NUCLEAR OPERATING COMPANY. INC.
DOCKET NO. 50-348 I
JOSEPH M. FARLEY NUCLEAR PLANT.' UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE j
J Amendment ~ No.112 License No. NPF-2 j
i 1.
The Nuclear Regulatory Comission_ (the Comission) has found that:
)
A.
The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated December 19, 1994, i
complies with the standards and requirements of the Atomic Energy l
Act of 1954, as amended (the Act), and the Comission's rules and
.l regulations set forth in 10 CFR Chapter I B.
The facility will operate in conformity with the application, the l
provisions of the Act, and the rules and regulations of the i
Commission; l
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations, 3
D.
The issuance of this license amendment will not be inimical to the' l
common defense and security or to the health and safety of the j
public; and E.
The issuance of this amendment is in accordance with 10 CFR Part i
I 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this' license amendment; and paragraph 2.C.(2) of Facility Operating License No.
j NPF-2 is hereby amended to read as follows:
i 9503060231 9 i
PDR ADOCK 0 348 P
e
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(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 112, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMISSION w-M -v --
William H. Bateman, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 1,1995
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ATTACHMENT TO LICENSE AMENDMENT NO.112 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Paaes Insert Paaes 3/4 7-2 3/4 7-2 3/4 7-3 3/4 7-3 B 3/4 7-1 B 3/4 7-1 B 3/4 7-2 B 3/4 7-2 1
i
i Table 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION 5
Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 1
87 4
1 2
48 3
28 i
i TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint i
Operating Steam Generator *
(Percent of RATED THERMAL POWER) 1 2
3
- At least two safety valves shall be OPERABLE on the non-operating steam generator.
- These values left blank pending NRC approval of 2 loop operation.
FARLEY-UNIT 1 3/4 7-2 AMENDMENT NO. 20.112
. _ ~.....
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TABLE 3.7-3 y
STEAM LINE VALVES PER LOOP K
h VALVE NWSER LIET SETTING * (13%)**
ORIFICE SIZE (SQ. IN.)
l zy a.
Q1N11VO - 10A, 11A, 12A 1075 psig 16 b.
Q1N11VO - 108, 11B, 12B 1088 psig 16 c.
Q1N11VO - 10C, 11C, 12C 1102 psig 16 d.
Q1N11VO - 10D, 11D, 12D 1115 psig 16 e.
Q1N11v0 - 10E, 11E, 12E 1129 psig 16 to N
A 4
0
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
- After testing, the valves will be left at ill.
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E 3/4.7 PLANT SYSTEMS EASES i
3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Editjon. The total relieving capacity for all valves on all of the stem 4 lines is at least 12,984,660 lbs/hr which is 112 percent of the total secondary steam flow of 11,613,849 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis.
l FARLEY-UNIT 1 B 3/4 7-1 AMENDMENT NO. 26.112
I J
l PLANT SYSTEMS EASES I
3/4.7.1.2 AUXILIARY TEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off-site power.
Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 330 gpm at a pressure of 1133 psia l to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1133 paia to the entrance of the steam generators. l 1
This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
j 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small 1
fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
)
This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the accident analyses.
l I
FARLEY-UNIT 1 B 3/4 7-2 AMENDMENT No. 26,112 4
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' **h UNITED STATES f
NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. anse6-4001
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SOUTHERN NUCLEAR OPERATING COMPANY. INC.
DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 103 License No. NPF-8 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated December 19, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the j
public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical
)
Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-8 is hereby amended to read as follows:
i.
.. -. e -.., _..._..
2-(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 103, are hereby incorporated in the license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION William H. Bateman, Director Project Directorate II-I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 1, 1995 l
l I
l
I J
ATTACHMENT TO LICENSE AMENDMENT NO. 103 TO FACILITY OPERATING LICENSE NO. NPF-8 DOCKET WO. 50-364 Replace the following pages.of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Paaes Insert Paaes 3/4 7-2 3/4 7-2 3/4 7-3 3/4 7-3 B 3/4 7-1 B 3/4 7-1 B 3/4 7-2 B 3/4 7-2 l
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4 Table 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON TLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION j
Maximum Number of Inoperable Maximum Allowable. Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 87 l
2 43 3
28 I
i TABLE 3.7-2 i
t".AXIMUM ALLOWABLE POWER RANGE NEUTRON TLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION j
Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 2
3 1
i
- At least two safety valves shall be OPERABLE on the non-operating steam generator.
- These values left blank pending NRC approval of 2 loop operation.
FARLEY-UNIT 2 3/4 7-2 AMENDMENT NO.
103
M h
TABLE 3.7-3
$l STEAM LINE VALVES PER LOOP Ki yALyg yunggR LIFT SETTING * (134)**
ORIFICE SIZE (SQ. IN. )
l h[j a.
Q2N11VO - 10A, 11A, 12A 1075 psig 16 b.
Q2N11VO - 10B, 118, 12B 1088 psig 16 c.
02N11VO - 10C, 11C, 12C 1102 psig 16 d.
Q2N11VO - 10D, 11D, 12D 1115 psig 16 e.
Q2N11VO - 10E, 11E, 12E 1129 psig 16 ta Na 4
d,
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
- After testing, the valves will be left at ill.
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3/4.7 PLANT SYSTEMS 1 BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 1106 (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving j
capacity is associated with a' turbine trip from 1006 NYTED THERMAL POWER coincident with an assumed loss of condenser heat sink (1.e., no steam bypass to the condenser).
l The specified valve lift settings and relieving capacities are in accordance with the requirements _of Section III of the ASME Boiler and l
I Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is at least 12,984,660 lbs/hr which is
{
112 percent of the total secondary steam flow of 11,613,849 lbs/hr at q
1006 RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per i
steam generator ensures that sufficient relieving capacity is available i
for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is all'owable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis.
f FARLEY-UNIT 2 3 3/4 7-1 AMENDMENT NO. 103
n-PLANT SYSTEMS BASES 3/4.7.1.2
' AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down.to less than 350*F from normal operating conditions in the event of a total loss of off-site power.
Each electric driven auxiliary feedwater pump is capable of s
delivering a total feedwater flow of 330 gym at a pressure of 1133 psia l to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1133 psia to the entrance of the steam generators. l This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.
1 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum j
water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the j
atmosphere concurrent with total loss of off-site power. The contained 1
water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the accident analyses.
i
)
103 FARLEY-UNIT 2 B 3/4 7-2 AMENDMENT NO.
l